ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Argonne National LaboratoryWest – Occupational Environmental Dose

Document Number:

ORAUT-TKBS-0026-4 

Effective Date: 09/30/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 26

Subject Expert: Henry Peterson

Document Owner

Approval: Signature on File Date: 09/27/2004

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Date: 09/27/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 09/28/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 09/30/2004 James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record Of Issue/Revisions ..................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

4.1 Introduction................................................................................................................................. 6

4.2 Internal Intakes from Onsite Airborne Radionuclide Concentration..............................................9

4.2.1 Operational Releases.........................................................................................................9

4.2.2 Episodic Releases at INEEL.............................................................................................13

4.2.2.1 SL-1 Reactor Accident.............................................................................................15

4.2.2.2 Criticality and Accident Occurrences at ICPP ..........................................................15

4.2.2.2.1 ICPP Criticality of October 16, 1959 ..............................................................16

4.2.2.2.2 ICPP Criticality of January 25, 1961 ..............................................................16

4.2.2.2.3 Fuel Element Cutting Facility Filter Break ......................................................17

4.2.2.3 Releases from Planned Tests..................................................................................17

4.2.2.3.1 Initial Engine Test 14 .....................................................................................17

4.2.2.3.2 Initial Engine Test 15(B) ................................................................................18

4.2.2.3.3 Initial Engine Test 17(B) ................................................................................18

4.2.2.3.4 Initial Engine Test 19(A) ................................................................................19

4.2.2.3.5 Initial Engine Test 25(A) ................................................................................19

4.2.2.3.6 Initial Engine Test 26(A) ................................................................................20

4.3 External Dose ...........................................................................................................................20

4.3.1 Facility Fence-Line Annual Doses ....................................................................................20

4.3.2 Facility Air Immersion Doses ............................................................................................22

4.4 Uncertainty ................................................................................................................................23
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 2 of 26

References ........................................................................................................................................24

LIST OF TABLES

Table

4-1 Intake by year for ANL, 1952-2002 ..........................................................................................

Page

11

4-2 Comparison of calculated facility intakes with intakes from environmental 

monitoring results ....................................................................................................................12

4-3 Intakes at EBR-I for criticalities at ICPP...................................................................................16

4-4 Intakes at EBR-I for FECF filter break incident at ICPP ...........................................................17

4-5 Intakes at EBR-I for Initial Engine Tests at INEEL ...................................................................18

4-6 INEEL facility fence direct gamma values (TLD – background) mR)......................................... 22

LIST OF FIGURES

Figure

4-1 INEEL site map........................................................................................................................

Page

7

4-2 Nine-year (1974 - 1983) average mesoscale dispersion isopleths of air 

concentration at ground level (hr2

/m3 x 10-9), normalized to unit release rate .........................10

4-3 Onsite and distant particulate beta concentrations in air ........................................................14

4-4 Dispersion coefficient contours for SL-1 accident...................................................................15

4-5 Example of onsite TLD Monitoring Data.................................................................................21
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 3 of 26

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 07/02/2004 00-A New Technical Basis Document for the Argonne 

National Laboratory – Occupational 

Environmental Dose. Initiated by Norman D.

Rohrig.

Draft 08/25/2004 00-B Incorporates NIOSH and internal review 

comments. Initiated by Norman D. Rohrig.

09/30/2004 09/30/2004 00 First approved issue. Initiated by Norman D. 

Rohrig.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 4 of 26

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AFSR Argonne Fast Source Reactor

ANL-W Argonne National Laboratory-West

ARA Auxiliary Reactor Area

ATR Advanced Test Reactor

BORAX Boiling Water Reactor Experiment

Bq Becquerel (1 disintegration per second)

CDC Centers for Disease Control and Prevention

CERT Controlled Environmental Release Test

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

CPP (Idaho) Chemical Processing Plant 

DOE U.S. Department of Energy

EBR-I Experimental Breeder Reactor No. 1

EBR-II Experimental Breeder Reactor No. 2

EFS Experimental Field Station

EMDR Environmental Monitoring Data Report

EMR Environmental Monitoring Report

ETR Engineering Test Reactor

F Fast (solubility rate)

fCi femtocurie

FECF Fuel Element Cutting Facility

FP Fission Product

FPFRT Fission Product Field Release Test

GE General Electric (Company)

GE-ANP General Electric-Advanced Nuclear Propulsion (Program)

H&S Health & Safety (management organization)

HTRE Heat Transfer Reactor Experiment

ICPP Idaho Chemical Processing Plant

IDO, ID Idaho Operations Office (of AEC)

IET Initial Engine Test

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INELHDE Idaho National Engineering Laboratory Historical Dose Evaluation

INTEC Idaho Nuclear Technology and Engineering Center

kW kilowatt

M Moderate (solubility rate)

mR milliRoentgen 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 5 of 26

mrem millirem

MTR Materials Testing Reactor

NCRP National Council on Radiation Protection and Measurements

NOAA National Oceanic and Atmospheric Administration

NRTS National Reactor Testing Station

RAC Radiological Assessment Corporation

RM radioactive material

RSAC Radiological Safety Analysis Computer (program)

RWMC Radioactive Waste Management Complex

S Slow (solubility rate)

SL-1 Stationary Low-Power Reactor No. 1

TAN Test Area North

TBD technical basis document

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor Test

U.S.C. United States Code

ZPR Zero Power Reactor

ZPPR Zero Power Plutonium Reactor
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 6 of 26

4.1 INTRODUCTION 

Technical Basis Documents (TBDs) and Site Profile Documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)). 

This TBD addresses radioactive material (RM) releases from areas or facilities at the Idaho National 

Engineering and Environmental Laboratory (INEEL), formerly the National Reactor Testing Station 

(NRTS) and later the Idaho National Engineering Laboratory (INEL), that could affect employees at 

the Argonne National Laboratory-West (ANL-W) facility. The analysis for this TBD divided releases 

into two components: (1) normal “chronic” operational releases, and (2) episodic releases that 

generally are of short duration. These releases potentially represent unrecorded or missed doses, as 

either direct gamma radiation or beta-gamma radiation from immersion in the radioactive gaseous 

cloud, for individuals who did not have personal dosimetry to record the dose, or as internal doses 

from RM inhalation.

In addition, this TBD addresses direct gamma doses resulting from facility operations. In general, 

these doses, if not controlled by management, increase with time and create a facility background 

dose. At INEEL, these facility background doses were recorded by film badges infrequently and 

inconsistently before 1970 and by thermoluminescent dosimeters (TLDs) on a routine basis since 

1972. These facility background doses, or facility fence-line doses, as they are sometimes called, are 

a nebulous indication of a dose that workers could receive if they inhabited outside areas within the 

facility. This TBD presents ANL-W facility fence-line doses (minus background) for three locations: 1) 

the EBR-I location for 1952 to 1972, 2) the Transient Reactor Test (TREAT) facility and the 

Experimental Breeder Reactor (EBR)-II both for 1972 to present. 

As outlined and discussed in Part 2 of the ANL-W facility profile (ORAU 2004), the U.S. Atomic 

Energy Commission (AEC) selected the INEEL Site as an isolated location for testing various reactor 

concepts. The INEEL is isolated from the public in two important aspects: (1) it is remote from 

population centers, and (2) it is isolated hydrologically because no surface streams originate on the 

Site and flow to an offsite location and no streams cross the Site. Although the INEEL sits above the 

large Snake River Aquifer that eventually surfaces and enters the Snake River in the Hagerman 

Valley area, the annual flow rate of the water in the aquifer is 5 to 15 feet per day.

ANL-W is a unique facility at the INEEL. Although inside the INEEL boundary, ANL-W is under the 

jurisdiction of the U.S. Department of Energy (DOE) Chicago Operations Office. Although the facility 

operates in accordance with 10 CFR 835, its operations have been and are more in line with a 

university atmosphere engaged with pure nuclear energy research; these operations support those of 

the University of Chicago or ANL-E, near Chicago, Illinois. During its first 14 years of operation (i.e., 

from 1951 through 1965), ANL-W was in the southwest corner of INEEL at the location of EBR-I. At 

this location ANL-W conducted EBR-I, Zero Power Reactor No. 3 (ZPR-III), Argonne Fast Source 

Reactor (AFSR), and all Boiling Water Reactor Experiments (BORAX) tests. In 1958, construction 

began on the TREAT Facility and on the EBR-II at the location presently designated as ANL-W at the 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 7 of 26

southeastern corner of INEEL. Since the mid-1970s, essentially all ANL-W operations have been 

conducted at the present ANL-W location, as shown on Figure 4-1.

Figure 4-1. INEEL site map. 

During the 50-year history of the INEEL Site, DOE and its predecessor agencies designed, built, and 

operated about 50 different reactor concepts at INEEL. All of these reactors have been prototype, 

low-power critical, or test reactors. INEEL operated no weapons production or commercial power 

reactors. Most, if not all, of these reactors have used highly enriched (93% or higher) uranium as fuel. 

Only a few have produced significant airborne effluent: (1) the Heat Transfer Reactor Experiment 

(HTRE) reactors, operated under the General Electric-Aircraft Nuclear Propulsion (GE-ANP) Program 

at the north end of the INEEL Site at Test Area North (TAN), (2) test reactors [Materials Testing 

Reactor (MTR), Engineering Test Reactor (ETR), and Advanced Test Reactor (ATR)], all at the Test 

Reactor Area (TRA) near the middle southern end of the Site, and (3) EBR-II at ANL-W at the 

southeastern corner of the Site, which has produced minor amounts of airborne effluent. 

Another historically important airborne effluent producer is the Idaho Nuclear Technology and 

Engineering Center (INTEC), formerly known as the Idaho Chemical Processing Plant (ICPP). This 

facility, constructed in the early 1950s, began processing nuclear fuel in February 1953 and continued 

SOUTH 

SECURITY GATE

▪

▪ ▪ CERRO GRANDE

FRENCHMAN’S

CABIN

▪

RENO RANCH
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 8 of 26

until 1992. Throughout its history, the “Chem” Plant, as it is commonly known, has reprocessed fuel 

from test reactors at INEEL, zirconium-clad fuel reclaimed from various reactors, stainless-clad fuel 

from EBR-II, and many AEC test reactors from around the world. Apart from the GE-ANP Program, 

which tested nuclear powered aircraft engine concepts with only one barrier (fuel cladding) between 

the fission products and the environment, TRA and ICPP airborne releases have been the most 

radiologically significant releases at INEEL (RAC 2002). Through the years that INEEL has published 

environmental monitoring reports, ICPP airborne effluents have been attributed to creating the 

majority of the INEEL “boundary dose.” Considering this fact, dose reconstructors should suspect 

ICPP airborne effluent as responsible for the maximum INEEL worker doses. Calculations performed 

for the INEEL TBD show that although ICPP airborne effluent is the most radiologically significant 

release at INEEL, the impact to all facility workers at INEEL is significantly below the allowable and 

acceptable limit. Figure 4-1 shows INEEL facility locations, including the EBR-I and ANL-W facilities.

From the beginning of operations at the INEEL Site, DOE and its predecessor agencies selected 

facility locations to limit the potential for operational releases at one facility to affect another facility. 

Because the Site encompasses 890 square miles, there was ample room to place facilities with this 

principle in mind. Because the Site has an average elevation of 5,000 feet and its general 

meteorological characterization indicates a nocturnal inversion from the north-northeast and a daytime 

lapse condition with winds from the southwest, transitional weather regimes are less frequent than at 

lower elevations. The 50-year history of the Site has demonstrated that the large expanse of INEEL 

and this meteorological characteristic have been effective in maintaining the operational isolation of 

each facility.

Beginning with the GE-ANP program, which began in the early 1950s, the INEEL Site has had the 

capability of plume tracking by aircraft. The local National Oceanic and Atmospheric Administration 

(NOAA) field office, which was dedicated to INEEL needs and requirements, provided plume 

projection capabilities for various programs with a rather extensive network of meteorological 

monitoring stations (Yanskey, Markee, and Richter 1966). The plumes from all intentional planned 

releases from the GE-ANP tests, the Controlled Environmental Release Test (CERT), the Fission 

Product Field Release Test (FPFRT), the Fuel Element Burn Tests A and B, etc., were directed over 

an instrumented monitoring grid (GRID III) that was remote from other facilities, such that releases did 

not affect other onsite facilities. 

INEEL reviewed and analyzed all airborne releases that have occurred since the beginning of Site 

operations as a result of a request from the DOE Idaho Operations Office (IDO) to evaluate the 

radiological impact to individuals at the INEEL boundary from airborne releases that had occurred 

since the beginning of Site operations. With the help of NOAA, which had hourly meteorological data 

from 1956 to that time, INEEL completed analyses for all airborne releases that occurred at the Site. 

Radiological consequences for an adult, a child, and an infant were calculated with Version 4 of the 

Radiological Safety Analysis Computer program RSAC-4 (Wenzel 1990). The results of the study 

were published in the Idaho National Engineering Laboratory Historical Dose Evaluation (DOE 1991); 

this TBD refers to that report as the INELHDE. All releases considered for that report are the bases 

for the releases considered in this TBD. In addition, all releases documented in the INELHDE, 

operational and episodic, have been independently reviewed and found, with minor modifications, to 

be substantially appropriate. The review, conducted by Radiological Assessment Corporation (RAC 

2002) at the request of the Centers for Disease Control and Prevention (CDC) and the State of Idaho, 

evaluated the methodology by which the RSAC-4 computer program performs dose calculations 

against the methodology favored by the National Council on Radiological Protection and 

Measurements (NCRP). It stated: “As a final point, Tables 7, 8, 9a, 9b, 10a, and 10b, and Figures 18

and 19 confirm that the NCRP method was suitable for these ranking purposes when the results are 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 9 of 26

compared with those using the RSAC code. In all cases, the RSAC code confirmed the results 

obtained using the NCRP methodology” (RAC 2002, p. 57). 

Version 6 of the RSAC code (Wenzel and Schrader 2001) is used extensively in this document to 

provide onsite concentrations due to episodic releases as well as other evaluations. For more 

information on the RSAC code, see Peterson (2004).

EBR-I was the first reactor to operate at the Site. It and BORAX-I through –V were in the 

southwestern corner of INEEL, operated under the AEC Chicago Operations Office by the University 

of Chicago as ANL-W. These low-power reactors produced essentially no radioactive airborne 

effluent. As the EBR-I and BORAX programs ended, ANL-W relocated the locations of EBR-II, the 

TREAT Facility, the Zero Power Plutonium Reactor (ZPPR), etc., to the eastern section of INEEL. 

The EBR-I location is now a historic landmark and is open to the public during the summer months.

All inhaled quantities and concentrations referred to in this TBD apply to individuals stationed at the 

ANL-W facility, at the South-West location through 1965, and at the South-East location beginning in 

1966. 

4.2 INTERNAL INTAKES FROM ONSITE AIRBORNE RADIONUCLIDE CONCENTRATION

This section addresses onsite concentrations of radionuclides and onsite internal personnel intakes 

from normal operational releases and from shorter term releases such as those from criticality 

incidents at ICPP and tests performed by the GE-ANP Program at TAN. As stated above, operational 

releases from ICPP and TRA have been the predominant and radiologically significant releases at 

INEEL during the history of the Site. For more discussion of these releases and their relationship to 

other, less significant releases, see Peterson (2004), DOE (1991), or RAC 2002. 

For worker dose reconstruction, the analyst should use default values for the calculation under 

consideration. When solubility is of concern, an insoluble oxide form for solids is recommended for 

analysis, with “S” and “M” type materials being the predominant form. Without more definitive 

information on the type of material, the dose reconstructor should use the material that maximizes the 

dose for a particular situation. When iodines are of concern, the dose reconstructor should consider 

them to be “F” type materials.

4.2.1

Estimation of onsite concentrations of radionuclides and resulting potential intakes from operational 

releases at INEEL facility locations employs the same methodology used to determine offsite 

concentrations for Site annual environmental monitoring reports. The release for each year of 

operation is exactly the same as that documented in INELHDE (DOE 1991) with one exception: an 

analysis to reduce the number of radionuclides and yet retain those that contributed about 95% of the 

inhalation dose (see Peterson 2004) reduced the number of radionuclides from 56 to 9 for the 

operational releases. 

Operational Releases

Meteorological dispersion factors applicable to each INEEL facility were picked from the annual 

average mesoscale dispersion isopleths of ground-level air concentrations as published in 

environmental monitoring reports, as described in INELHDE (DOE 1991). As described in Appendix 

B of that document, dispersion isopleths are available for years beginning in 1973, with the exception 

of 1978, when INEEL upgraded the telemetry system. For years prior to 1973, the TBD analysis used 

a 9-year average of mesoscale dispersion isopleths of ground-level air concentrations (Bowman 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 10 of 26

1984), shown in Figure 4-2. For 1993 to 2002, the analysis used annual average mesoscale isopleths 

from the annual environmental reports (ESRF 1994, 1995, 1996, 1997, 1998, 2000; Stoller 2002a,b,c) 

to calculate the facility annual concentration. 

Of the many facilities on INEEL, this analysis considered only the ANL-W facility. Yearly isopleth 

values for each ANL-W location (EBR-I for 1952 to 1965 and EBR-II for 1966 to present) have been 

extracted from the annual environmental monitoring reports and converted from the normalized 

annual concentration1 (hr2

/m3) to concentrations (Bq/m3) 

Figure 4-2. Nine-year (1974 - 1983) average mesoscale dispersion isopleths 

of air concentration at ground level (hr2

/m3 x 10-9), normalized to unit release 

rate (Bowman 1984). 

and multiplied by 2.4 × 103

 m3/yr (the amount of air breathed occupationally per year) to produce 

activity inhaled per year (Bq) for an occupational individual. Table 4-1 lists these values. 

The annual inhaled quantities (Bq/yr) listed in Table 4-1 are based on known and published INEEL 

annual airborne emissions. The following discussion provides information that is located in 

NRTS/INEL/INEEL documentation on facility environmental sampling/monitoring and data that can be 

compared with the calculated inhaled quantities of Table 4-1. “Environmental” air sampling at the 

facility areas has been performed at least since the mid-1950s where airborne effluents were known 

or suspected to exist. The early IDO Health and Safety (H&S) Division Annual Reports document 

many studies for defining radionuclide concentrations in the vicinity of different facilities. These 

 

1 As used at the INEEL, this quantity is the sum of 8,766 calculations of the hourly average χ/Q.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 11 of 26

Table 4-1. Intake (Bq yr-1) by year for ANL-W, 1952-2002. 

Nuclide Ce-144 I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90 Y-91

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1

1976 5.2E-5 3.2E-5 0.0E+0 8.1E-6 3.6E-6 8.1E-4 0.0E+0 3.9E-4 3.0E-2

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1

1986 2.9E-4 8.9E-5 0.0E+0 1.3E-6 9.7E-8 2.3E-3 0.0E+0 1.6E-5 4.3E-2

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0

1995 0.0E+0 2.1E-5 0.0E+0 3.3E-8 5.5E-9 0.0E+0 0.0E+0 2.3E-6 0.0E+0

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0

2000 0.0E+0 1.9E-3 0.0E+0 3.6E-5 3.6E-7 0.0E+0 0.0E+0 3.5E-3 0.0E+0

2001 0.0E+0 1.0E-3 0.0E+0 2.4E-7 2.9E-5 0.0E+0 0.0E+0 1.2E-4 0.0E+0

2002 0.0E+0 1.6E-4 0.0E+0 6.6E-6 1.8E-5 0.0E+0 0.0E+0 3.8E-3 0.0E+0
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 12 of 26

studies were specific for a given test, operation, or incident, however, and did not occur in a set 

location or for a standard duration. The 1963 Annual Progress Report of the IDO Health and Safety 

Division (AEC no date) contains some facility environmental monitoring data. INEEL developed a 

routine facility environmental monitoring program between 1963 and 1970. In 1968 and 1969, formal 

Environmental Monitoring Reports (EMRs) (AEC ‘1968’, AEC ‘1969’) reported alpha, beta, and I-131 

concentrations that can be correlated with the data of Table 4-1. The 1970 EMR (AEC ‘1970’) reports 

gross beta values measured at the Central Facilities Area (CFA) that can be correlated with Table 4-1 

(EBR-I) values. The analysis for this TBD reviewed EMRs between 1970 and 1990 for data that is 

available and usable for this correlation. 

Table 4-2 lists results of the comparison. Because of the large variation in measurements made, the 

ratio of values calculated from the EMRs (column 4) to that derived from releases shown in Table 4-1 

is not well behaved (geometric mean of 0.39 and geometric standard deviation of 5.6). Nevertheless 

this comparison provides confidence in the results of Table 4-1. The deviation of the average ratio 

from 1 is small compared to the default geometric standard deviation of 3 assumed for environmental 

results. 

Table 4-2. Comparison of calculated facility intakes with intakes from environmental 

monitoring results . 

Year Activity type

Average annual

Concentration

Annual inhaled

quantity (Bq)

Table 4 inhaled

quantity (Bq)

1963 β-γ 1.7E-11 μCi/cc 1510 1,310a

Pu-239 6.0E-16 μCi/cc 0.05 0.014a

1968 α 0.0022 pCi/m3 0.18 0.01a

β 0.64 pCi/m3 56 337a

I-131 <0.08 pCi/m3 <7.1 1.4a

1969 α 0.023 pCi/m3 2 2.4E-3a

β 2.95 pCi/m3 262 118a

I-131 0.123 pCi/m3 10.9 2.1a

1970 Gross β 6.0E-13 μCi/ml 53 74a

Max. gr. Β @ CFA 8.1E-13 μCi/ml 72 74a

1973 Gross β 95 +42 fCi/m3 8.4 0.8a

EBR-I Sr-90 3.4 fCi/m3 0.3 0.15b

Nb-95 1.0 fCi/m3 0.09 ---

Cs-137 7-17 fCi/m3 0.6-1.5 ---

Ce-144 4-8 fCi/m3 0.36-0.71 0.057b

EFS Sr-90 5.9 fCi/m3 0.52 0.15a

Nb-95 0.9-2.4 fCi/m3 0.08-0.2 ---

Ru-106 6-9.8 fCi/m3 0.53-0.87 0.27a

Cs-134 0.8-1.6 fCi/m3 0.07-0.14 ---

Cs-137 17-27 fCi/m3 1.5-2.4 ---

1976c Gross β 3-6E-14 μCi/ml 2.5-5 0.6-25d

1986 Kr-85 @ CFA 3.7E-11 μCi/ml 3,290 890e

1988 Kr-85 @ CFA 1.1E-10 μCi/ml 9,770 14,000e

1990 Kr-85 @ CFA 2.7E-11 μCi/ml 2,400 690e

a. Values from INEEL TBD Table 4-3 for CFA.

b. Values from INEEL TBD Table 4-5 for RWMC since EBR-I is near RWMC.

c. Of 90 monthly values (January through September) for 10 facility areas, 89 values were between 

3 × 10-14 and 6 × 10-14 μCi/ml.

d. Using the current tables with 11 radionuclides, the inhaled quantity is about 0.6 Bq; with the 

original tables with 44 radionuclides, the inhaled quantity is about 25 Bq. 

e. Values derived from tables in an earlier version of the TBD report that contained concentrations 

of all INEEL released radionuclides.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 13 of 26

Figure 4-3 shows the variation of INEEL Environmental Monitoring sampling results for the period 

from 1978 through 1986 which is typical for earlier years as well as for later years. This figure also 

shows the close correlation of environmental sample results acquired at “distant communities” and 

those acquired at INEEL facilities and the effect of foreign nuclear tests and the Chernobyl reactor 

accident on INEEL environmental sampling results. As shown on this figure, the INEEL average 

concentration has not differed from “distant community” concentrations by more than a factor of 2 for 

the 9-year period and is very similar for earlier and later periods; subsequent Environmental 

Monitoring Reports show the same correlation for the years before 1978 and after 1986. The greater 

perturbations in facility and distant community concentrations are nearly all correlated with fallout from 

nuclear tests. There was no discernible evidence that facility effluent or resuspension affected facility 

concentrations.

4.2.2

Of the 108 episodic releases (accidents and planned tests) analyzed in DOE (1991), only 16 had the 

potential to affect other INEEL facilities. Only 9 of the 16 events could have affected the EBR-I 

facility. Section 4.1.2.2 describes three of these events, two of the three criticalities at ICPP and the 

Fuel Element Cutting Facility Filter Break, which also occurred at ICPP. The other 6 were planned 

GE-ANP tests that could have affected the EBR-I facility, as follows:

Episodic Releases at INEEL

1. Initial Engine Test (IET) 14 4. IET 19(A)

2. IET 15(B) 5. IET 25(A)

3. IET 17(B) 6. IET 26(A)

For a given test, if there was an onsite facility between the point at which the test occurred and the 

affected offsite location, that test was conservatively assumed to have affected an onsite facility or 

facilities. For example, the FECF Filter Break occurred at ICPP and clearly contaminated an area 

south of ICPP. According to the meteorological dispersion at the time of the filter break, the affected 

offsite location was Frenchman’s Cabin. Because EBR-I is in the straight-line path between ICPP and 

Frenchman’s Cabin, a radiological impact analysis was conducted for the EBR-I. 

Because all other test releases listed above, which originated at the TAN facility, affected one location 

on the southern boundary [Frenchman’s Cabin (shown on Figure 4-1), as evaluated in DOE (1991)], 

they have been assumed to have affected EBR-I because that facility is on the plume trajectory from 

TAN. The following sections discuss these events. The SL-1 accident, which was widely publicized, 

is only included to show it did not affect any other facility at the INEEL. 

A concerted effort has been made to reduce the number of radionuclides involved in the releases for 

the episodic events. Overall, the mix of radionuclides for all the episodic events is complicated by the 

type (“aged” versus “fresh”), and the relative quantities of each. When viewed together, the episodic 

events can be categorized into three categories: criticalities that involve “fresh” fission products that 

have relatively short half-lives when compared to radionuclides released from the Fuel Element Burn 

Tests, for example, releases involving long half-lived, aged fission products (Fuel Element Burn Tests 

and the release from the Fuel Element Cutting Facility Filter Break), and releases from the remaining 

IET tests that released short half-lived radionuclides, which are generally characterized as “fresh” 

fission products, and long half-lived radionuclides, which are characterized as “aged” fission products. 

The latter category is unique to the GE-ANP Program because of the “direct-to-air” conversion nature 

of the tests. Therefore, within these categories, the number of radionuclides has been reduced to the 

number that preserves 95% of the original dose that was calculated for that particular location. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 14 of 26

Figure 4-3. Onsite and distant particulate beta concentrations in air.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 15 of 26

4.2.2.1 SL-1 Reactor Accident

One significant accident at INEEL in the last 51 years released substantial amounts of RM to the 

environment. On January 3, 1961, a steam explosion at the Stationary Low-Power Reactor No. 1 

(SL-1) facility [near the location of Auxiliary Reactor Area (ARA) II in Figure 4-1] killed three SL-1 

personnel and ruptured the SL-1 reactor vessel. This, in turn, propelled RM into the reactor building 

and then into the environment. The amount of the release and the path that the cloud traveled from 

the reactor building were carefully monitored and well documented (Gammill 1961; Horan and 

Gammill 1961; Kunze 1962). All radiological doses to personnel involved in the rescue and cleanup 

of the reactor building were carefully controlled and documented. 

The SL-1 accident did not affect any other INEEL facility with the effluent of RM. The effluent traveled 

to the south of the facility, as shown in Figure 4-4 (DOE 1991).

Figure 4-4. Dispersion coefficient contours for SL-1 accident (redrafted from Horan and Gammill 

1961).

4.2.2.2 Criticality and Accident Occurrences at ICPP

Three accidental criticalities have occurred at the ICPP (now INTEC). The first occurred on October 

16, 1959, the second on January 25, 1961, and the third on October 17, 1978. The 1978 criticality, 

which released essentially just noble fission gases produced during the criticality, occurred after ANLW operations had moved to the present South-East site location. The two earlier criticalities released 

RM during or shortly after the event; in both cases, the effluent was transported to the 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 16 of 26

south-southwest and potentially exposed personnel at the EBR-I facility. The Fuel Element Cutting 

Facility Filter Break accident also occurred at the ICPP and has been postulated to have impacted the 

EBR-I facility. Claimant favorable analyses, described below, defined the amount of potential 

radiological exposure that could have occurred to an individual at this location. 

4.2.2.2.1 ICPP Criticality of October 16, 1959

On October 16, 1959, at approximately 3:00 a.m., a criticality event occurred at the ICPP in the 

WH-100 vessel. The estimated magnitude of this event was no greater than 4 × 1019 fissions (DOE 

1991). Nuclear Incident at the Idaho Chemical Processing Plant (Ginkel et al. 1960) provides a full 

account of the incident and documents the radiological doses, calculated internal and measured 

external, for plant personnel involved in the incident.

For the calculation of intakes for this incident, meteorological conditions were modeled so the Χ/Q at 

22 km matched the value calculated for Frenchman’s Cabin (south of the INEEL, as shown on Figure 

4-1) where offsite doses were calculated and reported in DOE (1991). RSAC-6 was used to calculate 

Χ/Q values for EBR-I. These concentrations and intake quantities would be applicable only if the 

individual was in the respective areas on the morning of October 16, 1959. Table 4-3 lists intakes 

applicable at EBR-I. 

Table 4-3: Intakes (Bq/event) at EBR-I for 

criticalities at ICPP. 

Date 10/16/1959 1/25/1961

Event Criticality Criticality

Exposure location EBR-I Area EBR-I Area

Rb-89 2.1E+4 4.4E-1

Sr-91 2.2E+3 4.8E+0

Sr-92 2.6E+3 4.3E+0

Y-92 3.1E+2 2.2E+0

Y-93 2.4E+2 5.7E-1

Te-133

I-131(elem.) 1.2E+1 5.7E-2

I-133 2.6E+2 1.1E+0

I-134 1.9E+3 2.1E+0

I-135 8.2E+2 3.2E+0

Cs-138 3.1E+4

Ba-139 1.6E+4 1.9E+1

La-141 1.1E+3 3.0E+0

La-142 9.4E+2 1.2E+0

4.2.2.2.2 ICPP Criticality of January 25, 1961

The January 25, 1961 criticality occurred in ICPP vessel H-110 about 9:50 a.m. This event consisted 

of an estimated 6.0 × 1017 fissions. The report documenting the incident states:

Of the 251 individuals present in the ICPP area at the time of the incident, none 

received significant radiation exposure. The highest exposure as determined from film 

badge readings did not exceed 55 millirem of penetrating radiation. Essentially no beta 

radiation was detected. No significant neutron exposure or internal contamination from 

inhalation was found. The absence of significant exposures is attributable to the 

extensive shielding provided by the process cell in which the event took place and the 

control of the fission gases by the equipment. (Paulus et al 1961)
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 17 of 26

As for the 1959 criticality, Χ/Q values were calculated for EBR-I. The source term used for this event 

is the same as that used for DOE (1991). Table 4-3 lists the intakes applicable at EBR-I if the 

individual was in the area on January 25, 1961.

4.2.2.2.3 Fuel Element Cutting Facility Filter Break

The ends of fuel elements sent to ICPP contained structural components that were cut off before the 

elements were processed. Cutting these end pieces off and cutting the fuel elements into sections 

before they were sent to CPP-601 for processing occurred in the FECF in CPP-603. During the night 

of October 29 and early in the morning of October 30, 1958, INEEL conducted decontamination 

operations in the FECF. Acid fumes from the decontamination operations caused failure of the FECF 

exhaust filters, resulting in the release of particulate activity to the south of ICPP. 

Approximately 100 curies (Ci) of long half-life particulate RM was released over an area of 

approximately 200 acres (AEC 1959). The released radioactive material and quantities were the 

same as those published in DOE (1991). Table 4-4 lists the best-estimate intakes of the radionuclides 

in the EBR-I area. These intakes would be applicable only if the individual was in the area at the time 

of the release (i.e., during the night of October 29 and the early morning of October 30, 1958).

Table 4-4. Intakes (Bq/event) at EBR-I for FECF filter break incident at ICPP . 

Date Event

Exposure 

location Sr-89 Sr-90 Y-91 Zr-95 Ru-103 Ru-106 I-131 (elem.) Ce-144 Pr-143

10/29/58 

10/30/58

FECF Filter 

Break EBR-I Area 6.3E-1 6.3E-1 1.4E+0 2.0E+0 1.3E-1 6.6E-1 4.2E-12 9.0E+0 2.6E-6

4.2.2.3 Releases from Planned Tests

The following were all planned tests, conducted under the GE-ANP Program at Test Area North 

(TAN), which potentially affected the EBR-I facility. 

4.2.2.3.1 Initial Engine Test 14

IET 14 was the eighth nuclear test conducted by the GE-ANP program at TAN. This test was the fifth 

in the HTRE-2 reactor configuration. This test series involved the evaluation of the L2A-1 insert 

cartridge. The cartridge contained fueled and unfueled, beryllium-oxygen ceramic tubes. There was 

no coating on the inside surfaces of the fueled tubes (Pincock 1959).

A total of 100.25 hours was accumulated on the insert fuel cartridge at a maximum insert fuel 

temperature of approximately 2,500°F. The objectives of the test were to (1) evaluate the operational 

effect of water vapor corrosion on fueled beryllium-oxygen tubes operating at a constant reactor 

mixed mean discharge air temperature over approximately 100 hours, and (2) measure the fission 

product release rate from uncoated fueled tubes as a function of temperature and operating time 

(Pincock 1959). 

Table 4-5 lists the fission products released during the IET 14 test, and the intakes at EBR-I. An 

individual would have been exposed to these concentrations only if present at these locations 

between April 24 and May 19, 1959. The above intakes are for a total exposure period of 26 days.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 18 of 26

Table 4-5: Intakes (Bq/event) at EBR-I for Initial Engine Tests at INEEL . 

Period Test Exposure location Rb-89 Sr-89 I-131 (elem.) I-133 I-135 Cs-138 U-234

4/24 – 5/19/59 IET 14 EBR-I 5.8E+0 1.7E-2 2.1E+0 1.3E+1 1.8E+1 4.2E+1 3.6E-6

6/16-6/24/59 IET 15(B) EBR-I 3.3E-1 8.4E-4 4.5E-1 2.0E+0 3.2E+0 2.3E+0 3.9E-5

10/12-12/12/59 IET 17(B) EBR-I 2.2E-3 6.2E-3 5.7E-1 2.4E+0 2.5E+0 6.2E-1 5.5E-6

2/17 – 2/29/60 IET 19(A) EBR-I 1.3E-1 6.8E-3 9.9E-1 6.7E+0 9.9E+0 4.9E+0 4.2E-7

11/22-11/30/60 IET 25(A) EBR-I 7.1E-4 4.6E-4 3.6E-1 4.0E+0 5.1E+0 1.5E-1 1.5E-6

12/23-12/28/60 IET 26(A) EBR-I 1.7E+0 1.1E-2 1.8E+0 7.2E+0 1.1E+1 1.5E+1 2.5E-5

4.2.2.3.2 Initial Engine Test 15(B) 

IET 15 was conducted at TAN between May 27 and June 24, 1959. This test involved the evaluation 

of the L2C-1 insert cartridge, which was of the concentric ring design. The fuel sheet was made of a 

chromium-uranium dioxide-titanium core clad with an iron-chromium-yttrium alloy (Evans 1959). From 

this operation data was obtained to evaluate:

1. Endurance capabilities of the advanced metals at a design temperature of 2,000°F for 

extended periods (planned endurance testing to total 120 hours or more)

2. The structural and metallurgical integrity of the fuel sheet in this particular cartridge design

3. The nature and extent of fuel sheet damage, if any, and the effect on cartridge performance

4. The performance potentials of the cartridge

The operation was successfully conducted to accumulate 80.75 hours at an insert extrapolated fuel 

sheet temperature of 2,015°F. The operation was terminated after 80.75 hours due to a release of 

fission products of such a quantity as to indicate fuel sheet rupture of an extent sufficient to warrant 

inspection (Evans 1959).

The insert was visually examined after completion of testing. No damage had occurred to the outer 

fuel sheets of the cartridge; however, there were blisters on the inner fuel sheets. In some instances 

the blisters had ruptured. The FP release for this test was divided into two periods based on a review 

of effluent monitoring data. The period from June 3 to 15 was considered to be an operation before 

the development of significant fuel sheet blisters. June 16 to 24 comprised the second period, when 

effects of blistering were clearly observed. 

According to the meteorology of the testing period, the second period affected Frenchman’s Cabin. 

Accordingly, this analysis addressed the radiological impact on the EBR-I. Table 4-5 lists the release 

of fission products, which correspond with the Part B operation release documented in DOE (1991) for 

the intakes applicable at EBR-I. An individual would have been exposed to these concentrations and 

intakes only if present at the locations between June 16 and June 24, 1959. The above intakes are 

for the total 9-day exposure period.

4.2.2.3.3 Initial Engine Test 17(B) 

IET 17 occurred between October 12 and December 12, 1959. Releases of airborne radioactivity 

occurred between November 2 and December 12, 1959, when the reactor operated at power levels 

exceeding 100 kW. The test series involved the evaluation of the L2E-1 insert cartridge (Evans 

1960). Table 4-5 lists the intakes for an individual at EBR-I. An individual would have received these 

intakes only if present at the respective locations between November 2 and December 12, 1959. The 

intakes are for a total exposure period of 40 days.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 19 of 26

4.2.2.3.4 Initial Engine Test 19(A) 

IET 19, conducted between February 9 and April 30, 1960, was a test series in the HTRE No. 2 

reactor to evaluate the L2E-3 insert, which contained fueled and unfueled hexagonal beryllium-oxygen 

ceramic tubes. The tubes were coated on the inside with coextruded zirconia (zirconium dioxide) 

(Pincock 1960). The primary purposes for running the test were to:

1. Operate the L2E-3 fuel cartridge at peak temperatures of 2,500° F and 2,600° F for 

100 hours or more at each temperature level to evaluate the effectiveness of the 

zirconium-dioxide coating against hydrolysis and the release of fission products.

2. Operate the insert fuel cartridge at various temperature levels at specified intervals 

during the endurance testing to determine fission product release as a function of 

insert temperature.

3. Obtain additional information on the effectiveness of an electrostatic precipitator in 

removing fission products from the reactor effluent. (Pincock 1960).

Pincock (1960) summarized the estimated total fission product release for the test runs based on spot 

sampling and reported them as 10-minute-decayed curies. The total fission product release reported 

for IET 19 was 2,892 Ci. The release for this test was modeled as for DOE (1991). Table 4-5 lists 

intakes for EBR-I. An individual would have been exposed to these intakes only if present at the 

respective locations between February 17 and February 29, 1960. The intakes are for the total 

exposure period of 13 days.

4.2.2.3.5 Initial Engine Test 25(A) 

IET 25, performed between November 15 and December 19, 1960, was an extension of the Phase II 

testing program conducted in IET 18. The test was conducted in the HTRE No. 3 reactor 

configuration. Releases of airborne radioactivity corresponding to the significant periods of operation 

were assumed to have occurred between November 22 and December 15, 1960. The release at IET 

25(A) was assumed to have occurred from November 22 through November 30, 1960.

The purposes of test series IET No. 25 were to demonstrate the capabilities of the fuel 

elements above design temperatures and to confirm that the powerplant could achieve 

a full nuclear start as predicted. The reactor went critical on November 14, 1960, and 

the test program was completed on December 19, 1960. (Linn 1962).

Only the following summary of effluent monitoring activities and results was available: 

Continuous effluent monitoring was maintained to measure and record the activity 

released to the atmosphere by the powerplant. The maximum output was 3.4 

curies/hour (measured 10 minutes after release). The total output for the test series 

was 218 curies (measured 10 minutes after release). The maximum release rate for 

I-131 was approximately 0.7 curies/hour (measured 10 minutes after release). The 

total offsite inhaled and ingested dose was below measurable amounts during this test 

series. (Highberg et al. 1961)

For this analysis the release was modeled, as in DOE (1991), as a straight-line trajectory such that 

the centerline plume affected EBR-I. Table 4-5 lists the intakes for this test. An individual would have 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 20 of 26

received intakes only if present at the respective locations between November 22 and December 15, 

1960. The tabulated intakes are for a total exposure period of 24 days.

4.2.2.3.6 Initial Engine Test 26(A) 

IET 26, conducted in HTRE No. 2, occurred between December 22, 1960, and March 31, 1961 (Field 

1961). Releases of airborne activity for the total test were assumed to have occurred between 

December 23, 1960, and March 30, 1961, when the reactor operated at power levels exceeding 120 

kW. Releases for the IET 26(A) operation occurred from December 23 to 28, 1960. The insert under 

test was the L2E-6 cartridge, which consisted of fueled and nonfueled ceramic beryllium-oxide 

hexagonal tubes coated on the inner surface with zirconium dioxide.

The airborne release model was consistent with the model of DOE (1991) with an assumed 

straight-line trajectory between TAN and EBR-I. Table 4-5 lists the intakes for EBR-I. An individual 

would have been exposed to these intakes only if present at the respective locations between 

December 23 and December 28, 1960. The tabulated intakes are for a total exposure period of 6 

days.

4.3 EXTERNAL DOSE

External radiation dose at a facility can be created by direct radiation from two sources: direct 

beta/gamma radiation from the facility or airborne effluents released from the facility or from adjacent 

facilities. In general, direct beta/gamma radiation from the facility will increase with time because the 

general contamination of the area will increase. In addition, as a facility ages, radioactive sources 

tend to accumulate at the facility, causing the general background to increase with time. A 

responsible H&S organization will observe and curb such a trend to prevent personnel exposures 

from increasing unnecessarily. The following sections discuss facility fence-line film badge and TLD 

data that recorded doses from airborne fission product releases that had the potential for personnel 

exposure. Peterson (2004) contains more information on these two subjects.

4.3.1

Before 1970, many film badge or TLD measurements occurred inside the INEEL Site boundary. 

During the IET period at TAN (1956 to 1961), many film badges were placed along the highways that 

triangulated the IET area and along some of the highways at the southern end of the Site. Initially, the 

badges were retrieved and read once a month. The frequency changed to 6 weeks in 1962 and then 

changed back to monthly in 1963. Film badges were used through 9 months of 1966 and TLDs were 

used after that time. Beginning in 1967, TLDs were changed on a semiannual basis. Recorded 

significant readings during the film badge period showed that the maximum badge reading increased 

by only a factor of 2 or 3 above background. However, the location of the badge with the increased 

reading was not identified. Peterson (2004) contains more information and film badge data for this 

early period. The “detection limit” for the film badge reading was often quoted as 10 mrem for both 

beta and gamma readings (AEC 1963) and as 10 mrem for the TLD when it was first used. With the 

measured annual background radiation field at INEEL before operations began between 100 and 150 

mrem/year, the monthly value of 8 to 13 mrem is at the detection limit of the film badge or TLD. 

Therefore, the uncertainty for monthly changeouts is higher than for less frequent changeouts. 

Facility Fence-Line Annual Doses

Facility fence monitoring and facility locations were established between the latter part of 1970 and 

the latter part of 1972. No film badge or TLD information is available for the EBR I and BORAX 

locations when they were operational. From 1972 through 1983, facility fence TLD measurements, 

made on a 6-month basis with 5 TLDs at each facility position, are available in the Environmental 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 21 of 26

Monitoring Data Reports (EMDRs) for INEL. Figure 4-5 (Table II from ERDA 1976) shows that 

uncertainty can vary from less than 10% to 20% for a given set of readings. At each of the 34 listed 

monitoring locations, there were normally five TLDs for a potential of 170 readings for a 6-month 

period. For this particular 2-year set of data, 1.5% of the 136 sets of readings are assigned a 2 sigma 

uncertainty of 16% to 20%, and 18.4% of the readings are assigned a 2 sigma uncertainty of 11% to 

15%. However, 80% of the 136 values ascribed for the 34 locations over the 2-year period have a 2 

sigma uncertainty of less than 10%. 

Figure 4-5 Example of onsite TLD monitoring data. 

To supply facility values for the 1965-1972 period, the highest value from all subsequent years was 

used. Facility fence TLD measurements could not be located for 1984 through 1992, but for 1993 and 

beyond the INEL/INEEL EMRs include such facility fence-line measurements. For the 1984-1992 

period for which TLD measurements are missing, reasonable extrapolations provided the missing 

values. In addition, the EMRs recorded background TLD measurements corresponding to the facility 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 22 of 26

fence TLD measuring periods. Table 4-6 lists all reduced facility fence-line TLD data (facility fenceline data minus background) in the EMRs. Peterson (2004) contains a more detailed discussion of 

the data.

Table 4-6. INEEL facility fence direct gamma values (TLD – background) (mR).

Year EBR-II TREAT Background

1965-72 59 50 100-150

1973 37 19 121

1974 35 17 123

1975 32 8 118

1976 56 50 113

1977 22 0 132

1978 56 2 129

1979 59 5 113

1980 51 12 119

1981 28 9 118

1982 20 12 117

1983 24 10 115

1984 31 13 124

1985 31 13 124

1986 31 13 124

1987 31 13 124

1988 31 13 124

1989 31 13 124

1990 19 13 124

1991 19 13 124

1992 19 13 124

1993 28 16 111

1994 15 3 130

1995 17 7 116

1996 22 21 129

1997 16 16 128

1998 0 11 131

1999 13 13 122

2000 25 26 129

2001 0 3 140

2002 18 39 120

4.3.2

INEEL facility air-immersion (beta-gamma) doses could be calculated from the noble gas and halogen 

portions of the operational releases, and, if applicable, from the noble gas portion of the applicable 

episodic releases. This calculation should be unnecessary because these releases would be 

recorded in the fence-line TLD doses listed in Table 4-6.

Facility Air Immersion Doses

However, in considering this increased uncertainty, it is interesting to note facility air monitoring 

results that are also discussed in the annual EMRs. In each case, the facility air concentration is 

compared to that concentration for a distant community, usually Idaho Falls. Normally the 

concentration is indistinguishable from the concentration for the distant community, as discussed in 

Section 4.2.1. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 23 of 26

4.4 UNCERTAINTY

INELHDE (DOE 1991) contains a detailed discussion of the derivation of airborne releases for 

operational conditions and episodic events.

Operational Releases

Discussions with the INELHDE (DOE 1991) authors suggest that operational releases, which were 

monitored, could be low by a factor of not more than 2. When the annual normalized ground-level 

concentration values are applied to the operational releases, the uncertainty could increase.

Episodic Releases

As described in INELHDE (DOE 1991), the episodic releases are a maximum reasonable value, 

based on the amount of material available to be released, and the conditions of the respective test. 

For such a release, the inhaled quantities (in Bq) were maximized by assuming the downwind 

exposed individual was subjected to the plume centerline concentration for the total time, night and 

day in most cases, of the release. In spite of the original effort to be “reasonably conservative” in the 

exposure estimates, some of the authors stated that the release considered for a particular episodic

event might be low by as much as a factor of 3.

Film Badge and TLD Measurements

As discussed in Section 4.3.1, the uncertainty of individual measurements, made with film badges 

and TLDs, can be as high as +100%, depending on the frequency of changeout (i.e., once per month, 

which was generally the case with film badges). The data for 1965 to 1972 in Table 4-6 is based on 

the highest TLD 6-month values of 1972 for the respective facility. Although the GE-ANP IET tests 

were conducted in the late 1950s and early 1960s (the last IET, 26, ended on March 31, 1961), tests 

with planned releases were administratively and meteorologically controlled so the airborne effluent 

traveled to the northeast over the monitoring grid such that adjacent facilities were not affected. 

However, after 1967 when facility fence-line measurements were routinely made with TLDs, with five 

TLDs at a given location the uncertainty is generally ascribed at less than 10%. Occasionally, less 

than about 20% of the time, these measurements have an ascribed level of uncertainty as high as 

20%.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 24 of 26

REFERENCES

AEC (U.S. Atomic Energy Commission), 1959, 1958 Annual Report of the Health and Safety Division,

IDO-12012, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1963, Annual Progress Report 1962, Health & Safety 

Division, IDO-12033, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), no date, Annual Progress Report 1963, Health & Safety 

Division, IDO-12037, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), ‘1968’, Environmental Monitoring Report No. 23, July – 

December and Annual Summary – 1968, Idaho Operations Office, Health Services Laboratory, 

National Reactor Testing Station. 

AEC (U.S. Atomic Energy Commission), ‘1969’, Environmental Monitoring Report No. 25, July – 

December and Annual Summary – 1969, Idaho Operations Office, Health Services Laboratory, 

National Reactor Testing Station.

AEC (U.S. Atomic Energy Commission), ‘1970’, National Reactor Testing Station Environmental 

Monitoring Program Report, January – December 1970, Idaho Operations Office, Health 

Services Laboratory, Idaho Falls, Idaho.

Bowman, A. L., 1984, INEL Environmental Site Characterization Report, EGG-NPR-6688, 3 volumes.

Casto, W. R., 1980, “ICPP Criticality Event of October 17, 1978,” Nuclear Safety, 21, 5. 

DOE (U.S. Department of Energy), 1979, 1978 Environmental Monitoring Program Report for the 

Idaho National Engineering Laboratory Site, IDO-12082(78), Idaho Operations Office, Idaho 

Falls, Idaho.

DOE (U.S. Department of Energy), 1989, The Idaho National Engineering Laboratory Site 

Environmental Report for Calendar Year 1988, IDO-12082(88), Idaho Operations Office, Idaho 

Falls, Idaho.

DOE (U.S. Department of Energy), 1991, Idaho National Engineering Laboratory Historical Dose 

Evaluation, DOE/ID-12119, Volumes 1 and 2, Idaho Operations Office, Idaho Falls, Idaho.

ERDA (Energy Research and Development Administration), 1976, INEL Site Environmental 

Monitoring Data-First Quarter 1976, Idaho Operations Office, September

ESRF (Environmental Science and Research Foundation) 1994, The Idaho National Engineering 

Laboratory Site Environmental Report for Calendar Year 1993, DOE/ID-12082(93), Idaho 

Falls, Idaho.

ESRF (Environmental Science and Research Foundation) 1995, The Idaho National Engineering 

Laboratory Site Environmental Report for Calendar Year 1994, DOE/ID-12082(94), Idaho 

Falls, Idaho.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 25 of 26

ESRF (Environmental Science and Research Foundation) 1996, In Summary: Idaho National 

Engineering Laboratory Site Environmental Report for Calendar Year 1995, DOE/ID12082(95), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 1997, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 1996, DOE/ID12082(96), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 1998, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 1997, DOE/ID12082(97), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 2000, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 1998, DOE/ID12082(98), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 2001, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 1999, DOE/ID12082(99), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 2002, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 2000, DOE/ID12082(00), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 2002, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 2001, DOE/ID12082(01), Idaho Falls, Idaho.

ESRF (Environmental Science and Research Foundation, Inc.) 2002, Idaho National Engineering and 

Environmental Laboratory Site Environmental Report for Calendar Year 2002, DOE/ID12082(02), Idaho Falls, Idaho.

Evans, R. C., 1959, Operations Report, HTRE-2, L2C-1 Cartridge, XDC 59-11-179, November 16.

Evans, R. C., 1960, HTRE-2 Insert L2E-1 Cartridge Operations Report, XDC 60-3-111, March 8.

Field, J. H. (editor), 1961, Test Results from D101-L2E-6 Power Testing, IET #26, DC 61-6-704.

Gammill, W. P., 1961, “Environmental Aspects of the SL-1 Accident,” 7th Annual Meeting of the Pacific 

Northwest Section of the American Industrial Hygiene Association, September 22, 1961. 

Ginkel, W. L., C. W. Bills, A. O. Dodd, K. K. Kennedy, and F. H. Tingey, 1960, Nuclear Incident at the 

Idaho Chemical Processing Plant, IDO-10035, U.S. Department of Energy, Idaho Operations 

Office, Idaho Falls, Idaho.

Highberg, J. W., J. H. Field, K. J. Skow, C. B. Cannon, and C. Davidson, 1961, Preliminary Data 

Report IET Test Series No. 25 D102A2 Powerplant, DC 61-2-724, February 10.

Horan, J. R., and W. P. Gammill, 1961, “The Health Physics Aspects of the SL-1 Accident,” Health 

Physics Society Meeting, Las Vegas, Nevada, June 16, 1961. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0026-4 Page 26 of 26

Kunze, J. F., 1962, Additional Analysis of the SL-1 Excursion, Final Report of Progress, July through 

October 1962, IDO-19313, U.S. Department of Energy, Idaho Operations Office, Idaho Falls, 

Idaho. 

Linn, F. C., 1962, Heat Transfer Reactor Experiment No. 3, General Electric Direct-Air-Cycle Aircraft 

Nuclear Propulsion Program, APEX-906, June 15.

ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for the Argonne 

National Laboratory-West (ANL-W) – Site Description, ORAUT-TKBS-0026-2, Oak Ridge, 

Tennessee.

Paulus, R. C., A. O. Dodd, K. K. Kennedy, F. H. Tingey, and F. M. Warzel, 1961, Nuclear Incident at 

the Idaho Chemical Processing Plant, January 25, 1961, IDO-10036, U.S. Department of 

Energy, Idaho Operations Office, Idaho Falls, Idaho.

Pincock, G. D. (editor), 1959, Test Results from D101 l2a-1 Power Testing, IET #14, DC 59-8-728, 

August 19.

Pincock, G. D., 1960, Test Results from D101-L2E-3 Power Testing, IET #19, DC 60-7-717, July 8.

Peterson, H. K., 2004, Supporting Information for INEEL Environmental Dose TBD, February.

RAC (Radiological Assessment Corporation), 2002, Final Report, Identification and Prioritization of 

Radionuclide Releases from the Idaho National Engineering and Environmental Laboratory, 

Centers for Disease Control and Prevention, Department of Health and Human Services, 

Atlanta, Georgia.

Stoller (S. M. Stoller Corporation), 2002a, Idaho National Engineering and Environmental Laboratory 

Site Environmental Report Calendar Year 1999, DOE/ID-12082(99), Idaho Falls, Idaho.

Stoller (S. M. Stoller Corporation), 2002b, Idaho National Engineering and Environmental Laboratory 

Site Environmental Report Calendar Year 2000, DOE/ID-12082(00), Idaho Falls, Idaho.

Stoller (S. M. Stoller Corporation), 2002c, Idaho National Engineering and Environmental Laboratory 

Site Environmental Report Calendar Year 2001, DOE/ID-12082(01), Idaho Falls, Idaho.

Wenzel, D. R., 1990, Interim User’s Manual for RSAC-4, Radiological Safety Analysis Computer 

Program, Version 4.03, Westinghouse Idaho Nuclear Company, Inc.

Wenzel, D. R., and B. J. Schrader, 2001, The Radiological Safety Analysis Computer Program 

(RSAC-6) Users’ Manual, INEEL/EXT-01-00540, U.S. Department of Energy, Idaho National 

Engineering and Environmental Laboratory, Idaho Falls, Idaho.

Yanskey, G. R., E. H. Markee, Jr., and A. P. Richter, 1966, Climatography of the National Reactor 

Testing Station, IDO-12048, Idaho Operations Office, Idaho Falls, Idaho